View source for AEP-NRC-2015-86, Supplemental Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent With. Previously Licensed Conditions.

{{Adams
| number = ML15265A036
| issue date = 09/18/2015
| title = Supplemental Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent With. Previously Licensed Conditions.
| author name = Lies Q
| author affiliation = Indiana Michigan Power Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000315
| license number = 
| contact person = 
| case reference number = AEP-NRC-2015-86, TAC MF2916
| document type = Letter type:AEP, Response to Request for Additional Information (RAI)
| page count = 13
| project = TAC:MF2916
| stage = Supplement
}}

=Text=
{{#Wiki_filter:INDIA NA                                                                                      indiana Michigan Power MICHIGAN                                                                                      Cook Nuclear Plant POW!ER                                                                                        One Cook Place A unit of American Electric Power                                                              Indiana MichiganPower.com September 18, 2015                                                                        AEP-NRC-201 5-86 10 CFR 50.90 Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC, 20555-0001 Donald C. Cook Nuclear Plant Unit 1 Supplemental Response to "Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with. Previously Licensed Conditions (TAC No. MF2916)"

==References:==
: 1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Co'ok Nuclear Plant Unit 1 Docket No. 50-315, License Amendment Request Regarding Restoration of Normal Reactor Coolant System Operating Pressure and Temperature Consistent with Previously Licensed Conditions," dated October 8, 2013, Agencywide Documents Access and Management System (ADAMS) Accession Number ML13283A121.
: 2. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit 1, Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure And Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)," dated March 31, 2014, ADAMS Accession Number ML14066A311.
: 3. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, ".Donald C. Cook Nuclear Plant Unit 1, Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)'," dated April 29, 2014, ADAMS Accession Number ML14121A422.
: 4. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit I     - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)," dated May 6, 2014, ADAMS Accession Number ML14099A450.
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U. S. Nuclear Regulatory Commission                                                AEP-NRC-201 5-86 Page 2
: 5. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)'," dated June 5, 2014, ADAMS Accession Numbers ML14181A537 and ML14181A538.
: 6. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)'," dated July 3, 2014, ADAMS Accession Number ML14189A102.
: 7. Email from M. L. Chawla, NRC, to H. L. Etheridge, I&M, "Request for Additional Information - Round 2 by Containment and Ventilation Branch Donald C. Cook Unit I Technical Specification Changes in Reactor Coolant System Normal Operating Pressure and Temperature Docket No. 50-31 5," dated August 8, 2014.
: 8. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, "Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)'," dated September 30, 2014, ADAMS Accession Number ML14275A453.
: 9. Letter from M. L. Chawla, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit 1 - Nuclear Regulatory Commission Staff Review of Licensee Response to Request for Additional Information Related to a License Amendment Request to Restore Normal Operating Pressure and Temperature (TAC No. MF2916)," dated October 22, 2014, ADAMS Accession Number ML14283A271.
By letter dated October 8, 2013 (Reference 1), Indiana Michigan Power Company (I&M) submitted an. application .for a license amendment to restore the normal reactor coolant system operating pressure and temperature consistent with previously licensed conditions for the Donald C. Cook Nuclear Plant, Unit 1. The U. S. Nuclear Regulatory Commission (NRC) staff provided a Request for Additional Information (RAI) (Reference 2) to complete the review of Reference 1. I&M responded to Reference 2 by Reference 3. By letter dated May 6, 2014, the NRC provided an additional RAI (Reference 4). to complete the review of Reference 1. l&M responded to Reference 4 by References 5 and 6. The NRC provided an additional RAI (Reference 7) to complete the review of Reference 1. I&M responded to Reference 7 by Reference 8. In Reference 8, I&M stated that a supplemental response to Reference 7 would be provided to the NRC at a future date. An NRC acknowledgement to Reference 8 was sent to I&M by Reference 9.
This letter provides I&M's supplemental response to the RAI in Reference 7. to this letter provides an affirmation statement. Enclosure 2 to this letter provides I&M's supplemental response to RAI SCVB-RAI-10 in Reference 7. Enclosure 3 to this letter provides the regulatory commitment made in this supplemental response.

U. S. Nuclear Regulatory Commission                                     AEP-NRC-201 5-86 Page 3 Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely Q./Shne Lies Engineering Vice President, Indiana Michigan Power JMT/ams

==Enclosures:==
: 1. Affirmation
: 2. Supplemental Response to SCVB-RAI-1 0
: 3. Regulatory Commitments c: A. W. Dietrich, NRC Washington DC J. T. King, MPSC MDEQ    - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures

ENCLOSURE 1 TO AEP-NRC-2015-86 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am Engineering Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf, of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.
Indiana Michigan Power Company haeLies Engineering Vice President, Indiana Michigan Power SWORN TO AND SUBSCRIBED BEFORE ME THIS           DAY OF ~-...x<r'-\*,*      ,2015 Notary DANIELLE  BURGOYNE Public, State of Michigan County of Berrien My Commission Expires 04-04-2018" Acting In the County of           ',.-,
MysCo~mmission Expires     c* \ -t*      - o**

ENCLOSURE 2 TO AEP-NRC-2015-86 Supplemental Response to SCVB-RAI-10 Introduction By letter dated October 8, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML13283A121), Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1, submitted a License Amendment Request (LAR). I&M proposed to increase the Reactor Coolant System Normal Operating Pressure and Temperature (NOP/NOT) consistent with previously licensed operating pressure of 2250 pounds per square inch absolute and full power average temperature of 571 degrees Fahrenheit (0F), to mitigate the ongoing Steam Generator (SG) tube wear.
I&M's letter dated October 8, 2013, was supplemented by letters dated April 29, 2014 (ADAMS Accession Number ML14121A422), June 5, 2014 (ADAMS Accession Numbers ML14181A537 and ML14181A538), July 3, 2014 (ADAMS Accession Number ML14189A112), and September 30, 2014 (ADAMS Accession Number ML14275A453), which provided responses to the U. S. Nuclear Regulatory Commission's (NRC) previous Requests for Additional Information (RAls). In the letter dated September 30, 2014, l&M stated that a supplemental response would be provided to the NRC RAI SCVB-RAI-10 following completion of additional analysis. This Enclosure provides I&M's supplemental response to SCVB-RAI-10.
By e-mail dated August 8, 2014, the NRC provided an additional RAI to I&M. The NRC RAI contained Westinghouse proprietary information which was 'identified by the NRC with underlining the information and enclosed in double square brackets. ((This sentence is an exampele] In order to withhold this information from public disclosure, the wording from the RAI has been redacted and enclosed in brackets, similar to the RAI. This is an example:
The RAl text is restated below. For completeness, the RAI is followed by I&M's first response from letter dated September 30, 2014. I&M's first response to the RAI is followed by I&M's supplemental response to the RAI.
SCVB-RAI-1 0 I]
applies to all containment types such as CNP Unit 1 which has an ice condenser having design pressure 12 psig smaller compared to a typical large dry PWR containment having a design pressure 55 psig. It would be unreasonableto assume that [L               J]   would apply to all containment types because it would imply that WCAP-10325-P-A along with WCAP-8354-P-A LOTIC methodology would over predict about 100% for CNP, and about 12% for a typical large dry containment. Therefore the NRC staff considers the applicabilityof ((            J]    in the peak containment pressure unacceptable without validation by performing a CNP Unit 1 specific containment pressure response analysis.
to AEP-NRC-2015-86                                                           Pg Page 2 Response to SCVB-RAI-3(c) reports that the LOCA peak containment pressure is 12.46 psig which exceeds the design pressure of 12 psig. The response recommends the following two independent actions to mitigate the specific 0. 72 psig impact associated with assuming higher initial containment temperatures.
o   Decreasing the maximum assumed Residual Heat Removal (RHR) containment spray actuation time from 4500 seconds to 4200 seconds, which provides a benefit of approximately one psi OR o    Crediting additionalice in the ice condenser above the TS minimum value; crediting an additionalice mass of 25,000 pounds-mass provides a benefit of approximately 0. 7 psi A CNP Unit 1 plant specific LOCA containment pressure response analysis is necessary.
Without the analysis, the estimated peak containment pressure is approximately 15.5 psig which is unacceptable because it is higher than the containment design pressure of 12 psig. The following table summarizes the breakdown of the 15.5 psig           -  estimated value of the containment peak pressure:
Source Document                                                          Peak Containment
____________________________________________Pressure Correction to LAR value of 11.7184 psig, as reported in response to      1.77p SCVB-RAI=-3(c)
Penalty due to non-conservative initial temperature, response to         07     s SCVB-RAI-3(c)
NSAL-11-5 (Table 1) penalty                                              2.8 psi NSAL- 14-2 (under heading "Technical Evaluation') penalty                0. 2 psi minimum NSA L-06-6 (under heading "Safety Significance')penalty                  5.8tps1ppoimtl Total (excluding the NSA L-06-6 penalty)                                15.45 7 psig Note 1: For ice condenser containment, NSAL-06-6 states the "Instead of applying the impact in a pressure increase, the penalty was converted into an energy value." The real value of penalty is therefore unknown.
Please revise the CNP Unit 1 LOCA AOR incorporating the above recommendations. In performing the revised licensing basis LOCA analysis, use the corrected WCAP-1O325-P-A methodology which has the removed the errors reported in NSA L-06-6, NSAL-1 1-5, and NSA L-14-2 and any other known errors. Confirm that based on the new licensing basis analysis, the peak containment pressure is less than the containment design pressure of
*12 psig. Provide the containment pressure profile and the peak pressure value for the limiting design basis LOCA.
I&M's First Response to SCVB-RAI-10 SCVB-RAI-10 requests that I&M reanalyze the loss of coolant accident (LOCA) mass and energy (M/E) release and containment analysis utilizing the currently licensed WCAP-10325 (Reference 1) methodology with the errors from NSAL-06-6 (Reference 2), NSAL-1 1-5 (Reference 3), and NSAL-14-2 (Reference 4) corrected.       The WCAP-1 0325 methodology     is a to AEP-NRC-2015-86                                                             Pg Page 3 non-mechanistic calculation of the LOCA M/E releases that also contains overly-conservative assumptions inherent to the methodology and over-predicts the CNP peak calculated containment pressure. With the errors in the WCAP-1 0325 methodology corrected, the results of the analysis are expected to exceed the 12 pounds per square inch gauge (psig) design limit for CNP and; therefore, reanalysis utilizing the current methodology is not considered an acceptable approach. Due to the limitations of the WCAP-10325 methodology, l&M intends to reanalyze the LOCA M/E release and containment response analysis utilizing the Westinghouse WCAP-1 7721-P (Reference 5) methodology which is currently under review by the NRC. The WCAP-17721-P methodology provides a mechanistic calculation of the M/E releases and is expected to result in calculated pressure below CNP's design pressure, based on initial results provided by Westinghouse to I&M. l&M plans to adopt the WCAP-17721-P methodology after NRC review and approval of the generic topical report, which is anticipated to be in March of 2015. I&M plans to provide, as a supplemental response to this RAI, the results of the revised analysis and appropriate supporting information within three months of NRC approval of the generic topical report.
l&M's previous responses to the RAIs indicated that CNP was relying on the generic margin, identified in Reference 3 for WCAP-10325, for continued operability.             As discussed in SCVB-RAI-10, the generic analysis margin was not stated to apply for all designs. Based on the undefined applicability of the generic margin, I&M has revised the justification for continued operability with a plant specific calculation using the unapproved WCAP-1 7721-P methodology.
This plant specific operability analysis was implemented consistent with the guidance in the NRC Inspection Manual Chapter 0326, Section C.04, and is documented in the CNP corrective action program. The WCAP-1 7721-P CNP analysis provides the basis for continued operability and therefore, the generic margin is not required. The supplemental WCAP-10325 operability analysis crediting additional ice, identified in I&M's previous response to the NRC's SCVB-RAI-5b, is still applicable under this new approach.

==References:==
: 1. WCAP-1 0325-P-A (Proprietary), 'Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version," May 1983.
: 2. NSAL-06-6, "LOCA Mass and Energy Release Analysis," June 6, 2006.
: 3. NSAL-1 1-5, "Westinghouse LOCA Mass and Energy Release Calculation Issues," July 25, 2011.
: 4. NSAL-14-2, "Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties," March 31, 2014.
: 5. WCAP-17721-P, "Westinghouse Containment Analysis Methodology         - PWR LOCA Mass and Energy Release Calculation Methodology."
I&M's Supplemental Response to SCVB-RAI-10 Consistent with the I&M commitment in the first response to SCVB-RAI-10 in Reference 1 and the subsequently issued NRC acknowledgement Letter (Reference 2), CNP has performed a to AEP-NRC-2015-86                                                           Pg Page 4 new Unit 1 LOCA M/E release calculation using WCAP-17721-P methodology (Reference 3) and an associated containment response using WCAP-8354 (Reference 4) under conditions that bound Unit 1 NOP/NOT. The LOCA M/E release calculation and containment response analysis will be implemented into the CNP licensing basis utilizing the 10 CFR 50.59 process.
The CNP Unit 1 analysis will be implemented into the CNP license basis by November 20, 2015.
The purposes of this Supplemental Response are to describe the new CNP Unit 1 containment integrity analysis, address the applicability of previous errors reported in NSAL-06-6 (Reference 6), NSAL-1 1-5 (Reference 7), and NSAL-14-2 (Reference 8) and any other known errors to the methodologies used in the new analysis, and present results [containment pressure profile and the peak pressure value for the limiting design basis LOCA] demonstrating that the peak containment pressure is less than the containment design pressure of 12 psig.
Description of New LOCA M/E Release Calculation and Containment Response Analysis The new analysis bounds the operation of Unit 1 as described in the LAR submitted in Reference 11 for the proposed return of the unit to NOP/NOT conditions. Full double-ended ruptures were analyzed in the cold leg and the pump suction leg, each with one and two trains of safety injection assumed, to cover the spectrum of possible limiting break locations for CNP Unit 1 in the context of the new methodology. The CNP Unit 1 LOCA M/E release calculation using the WCAP-17721 (Reference 3) methodology was performed without exception to the NRC approved methodology documented in Reference 5. The LOTICI code (Reference 4) was used to determine the containment response for each case, which is not a change from the current CNP license basis methodology.
The analysis uses an ice condenser ice mass of 2.2 x 106 pounds mass, which is consistent with current Technical Specifications (TS). It also uses a maximum residual heat removal (RHR) spray time of 4200 seconds that is consistent with the previous response to SCVB-RAI-10 in Reference 12 and reflects the formal change in the maximum RHR spray actuation time implemented at CNP for Unit 1 using the 10 CFR 50.59 process in 2014.
Consistent with the NRC Safety Evaluation (Reference 5), a conservative initial accumulator pressure was modeled which bounds plant operating conditions and accounts for measurement uncertainty. Finally, a sensitivity evaluation of initial ambient containment temperature conditions was performed to determine if the maximum or minimum temperatures are conservative with respect to containment integrity. Based on information obtained from the sensitivity, a bounding set of initial conditions was used in the analysis.
Applicability of Known Errors in New LOCA M/E Release Calculation and Containment Response Analysis NSAL-06-6 The generic issues described in NSAL-06-6 (Reference 6) are unique to the LOCA M/E analysis methodologies in WCAP-8264-P-A Rev. 1 (Reference 9) and WCAP-10325 (Reference 10) series of codes. It should be noted that Westinghouse informed CNP that the errors identified in NSAL-06-6 were not applicable to the CNP-specific WCAP-1 0325 analysis. Additionally, the errors reported in Reference 6 have no effect on LOCA M/E release calculation performed in accordance with the WCAP-17721-P to AEP-NRC-2015-86                                                           Pg Page 5 (Reference 3) methodology. The LOCA M/E release calculation using References 9 and 10, and the LOCA M/E release calculation using Reference 3 both are used in the same containment response analysis methodology (Reference 4) which is not impacted by Reference 6.
NSAL-1 1-5 NSAL-11-5 (Reference 7) is applicable to LOCA M/E release calculations performed for Westinghouse-designed pressurized water reactors utilizing the methodologies documented in WCAP-10325-P-A (Reference 10) and WCAP-8264-P-A, Revision I (Reference 9) as specified in Reference 7. NSAL-1 1-5 identified four errors that applied to CNP. The most significant being an error in the EPITOME code, which caused an underestimation of the calculated LOCA M/E releases. The code issues identified in NSAL-1 1-5 do not affect the new Unit 1 LOCA M/E release calculation and containment response analysis that use the methodologies described in WCAP-17721-P (Reference 3) and in WCAP-8354 (Reference 4) because the code containing the errors is not used.
NSAL-I14-2 NSAL-14-2 (Reference 8) identified errors in the LOCA M/E release calculations related to the modeling of the specific heat capacity of the SG thick metal mass in WCAP-8264-P-A, Rev.1 and WCAP-10325-P-A (References 9 and 10, respectively).
The WCAP-1 0325 (Reference 10) methodology used in the CNP analysis of record (AOR) utilized a non-conservative value for density and specific heat capacity of the SG thick metal mass.       The WCAP-17721-P (Reference 3) methodology utilizes the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code values for specific heat capacities as stipulated in the NRC Safety Evaluation (Reference 5). As a result, the new CNP Unit I LOCA M/E release calculation and containment response analysis are not impacted by NSAL-14-2.
Initial Containment Temperature I&M previously notified the NRC in Reference 12 of an error in the initial containment temperatures assumed in the containment analysis performed for Unit 1 NOP/NOT. To address this issue in the new Unit 1 LOCA M/E release calculation and containment response analysis, a sensitivity evaluation was performed utilizing the WCAP-17721-P (Reference 3) M/E releases along with the existing Reference 4 LOTIC1 containment methodology. The timing of ice bed melt-out relative to the initiation of RHR spray dictates when minimum or maximum temperatures are conservative. The evaluation ensured that the most limiting set of initial conditions for the ambient temperatures in upper and lower containment was used.
Initial Ice Bed Temperature Westinghouse previously notified I&M of an issue related to the initial ice bed temperature modeled in the LOTIC computer code used in the WCAP-8354 containment analysis methodology.       TS Surveillance Requirement 3.6.11.1 for ice condenser containment designs limits the maximum allowable temperature in the ice bed to    < 27&deg;F.
For current AOR which uses the WCAP-1 0325 M/E releases, the initial temperature of the ice bed was modeled at 15&deg;F rather than the TS maximum value. Increasing the initial ice bed temperature reduces the effectiveness of the ice condenser and results in to AEP-NRC-2015-86                                                            Pg Page 6 a peak pressure penalty. The WCAP-8354 containment analysis methodology is still used with the new analysis which uses the WCAP-17721 M/E releases; however, the initial ice bed temperature was corrected to the TS maximum value and therefore a limiting initial ice bed temperature is used in the new Unit 1 analysis and reflected in Figure 1.
InfoGram 14-1 Westinghouse issued InfoGram 14-1 (Reference 13) due to issues with the volumetric heat capacity of stainless steel utilized in the WCAP-1 0325 methodology.            The WCAP-10325 methodology uses a lower volumetric heat capacity than currently listed in the ASME Boiler and Pressure Vessel Code and would under calculate the metal energy released during the transient. A similar concern was raised by the NRC during the NRC's review of the WCAP-17721 Topical Report and is documented as a limitation in the NRC Safety Evaluation (Reference 5).          Similar to the response provided for NSAL-14-2, the CNP LOCA M/E releases calculated using the WCAP-17721 methodology utilize the ASME Boiler and Pressure Vessel Code values for volumetric heat capacities to ensure conservative modeling of metal energy is released during the transient.
No other known errors exist in the LOCA M/E release calculation or containment~ response analysis or their corresponding methodologies that could affect the results of the new CNP Unit 1 analysis described in this supplemental response to SCVB-RAI-10.
Analysis Results The calculated peak pressure from the limiting set of conditions for the CNP Unit I LOCA M/E release calculation and containment response analysis is 10.37 psig, which includes the 0.27 pounds per square inch (psi) containment pressure Updated Final Safety Analysis Report rack-up identified previously in Reference 12. Figure 1 below contains the direct output from the analysis and does not include the 0.27 psi rack-up. The results of the new containment integrity analysis are acceptable because they are less than the design pressure limit of 12.0 psig. With the implementation of the Reference 3 LOCA M/E analysis, all known outstanding issues with the AOR for Unit 1 will be resolved. Figure 1 below provides the Unit 1 peak pressure transient plot for the limiting WCAP-17721-P (Reference 3) M/E releases utilizing the current license basis, containment analysis methodology (Reference 4).
to AEP-NRC-201 5-86                                                        Pg Page 7 PSYS               45           0         0  Gouge   Pressure 10.5"1_
03 03 0                5000                 10000             15000              20000 Time     (s)
Figure 1: Unit 1 Peak Containment Pressure References
: 1. Letter AEP-NRC-2014-75 from J. P. Gebbie, Indiana Michigan Power Company (I&M),
to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)',"
dated September 30, 2014, Agencywide Documents Access and Management System (ADAMS) Accession Number ML14275A453.
: 2. Letter from Mahesh Chawla, NRC, "Donald C..Cook Nuclear Plant, Unit 1          - Nuclear Regulatory Commission Staff Review of Licensee Response to Request for Additional Information Related to a License Amendment Request to Restore Normal Operating Pressure and Temperature (TACF NO. MF2916)," of October 22, 2014, ADAMS Accession No. ML14283A271.
: 3. WCAP-17721-P, Westinghouse Containment Analysis Methodology         -  PWR LOCA Mass and Energy Release Calculation Methodology," April 2013.
to AEP-NRC-201 5-86                                                        Pg Page 8
: 4. WCAP-8354-P-A, "Long Term Ice Condenser Containment Code               -  LOTIC Code,"
April 1976.
: 5. "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-1 7721-P," Revision 0, and WCAP-17721-NP, Revision 0, "Westinghouse Containment Analysis Methodology - PWR [Pressurized Water Reactor]
LOCA [Loss-of-Coolant Accident] Mass and Energy Release Calculation Methodology,"
August 24, 2015 ADAMS Accession No. ML15221A005.
: 6. NSAL-06-6, "LOCA Mass and Energy Release Analysis," June 6, 2006.
: 7. NSAL-1 1-5, "Westinghouse LOCA Mass and Energy Release Calculation Issues,"
July 25, 2011.
: 8. NSAL-14-2, "Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties," March 31, 2014.
: 9. WCAP-8264-P-A, Revision 1, "Topical Report         -  Westinghouse Mass and Energy Release Data for Containment Design," August 31, 1975.
: 10. WCAP-10325-P-A, "Westinghouse LOCA Mass and               Energy Release     Model for Containment Design March 1979 Version," May 1, 1983.
: 11. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315, License Amendment Request Regarding Restoration of Normal Reactor Coolant System Operating Pressure and Temperature Consistent with Previously Licensed Conditions," dated October 8, 2013, ADAMS Accession Number ML13283A121.
: 12. Letter AEP-NRC-2014-51 from Q. S. Lies, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916),'
dated May 6, 2014," July 3, 2014, ADAMS Accession Number ML14189A102.
: 13. InfoGram 14-1, "Material Properties for Loss-of-Coolant Accident Mass and Energy Release Analyses," November 5, 2014.

Enclosure 3 to AEP-NRC-2015-86 REGULATORY COMMITMENTS The following table identifies an action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments. All commitments discussed in this table are one-time commitments.
Commitment                                Scheduled Completion Date (if applicable)
The CNP Unit I analysis for containment pressure, which is based on WCAP-17721-P and the associated NRC Safety Evaluation                  Noebr2,01 dated August 24, 2015, will be implemented into the Donald C. Cook Nuclear Plant license basis.}}